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The Japan Atomic Power Company is a company initially formed to jump start the commercial use of nuclear power in Japan, and currently operates two different sites. According to the official web site, JAPC is "the only power company in Japan solely engaged in nuclear energy".JAPC owns both units at the Tōkai Nuclear Power Plant and the Tsuruga Nuclear Power Plant with plans to expand at Tsuruga.The company is jointly owned by Japan's major electric utilities: The Tokyo Electric Power Company , Kansai Electric Power , Chubu Electric Power , Hokuriku Electric Power Company , Tohoku Electric Power , and Electric Power Development Company . Wikipedia.


Sugino W.,Japan Atomic Power Company
Nihon Kikai Gakkai Ronbunshu, B Hen/Transactions of the Japan Society of Mechanical Engineers, Part B | Year: 2013

Flow Accelerated Corrosion (FAC) of carbon steel (CS) piping is one of main issues in secondary system of Pressurized Water Reactor (PWR) nuclear power plant. Therefore, Oxygenated Water Chemistry (OWC) as a new approach to FAC suppression for PWR secondary system was applied to condensate system of Tsuruga-2 (1160 MWe PWR, commercial operation started in 1987) in Jan. 2011. To evaluate the FAC mitigation effect of OWC, wall thickness of actual condensate piping in Tsuruga-2 after OWC application was measured by continuous monitoring system, using high-temperature and high-resolution ultrasonic probe. As a result, it was demonstrated that FAC was mitigated by more than 5ppb of oxygen under Low-AVT (pH9.3) condition, and FAC was almost stopped even in 2ppb of oxygen under High-AVT (pH9.8) condition. © 2013 The Japan Society of Mechanical Engineers. Source


Murakami H.,Japan Atomic Power Company
Zairyo to Kankyo/ Corrosion Engineering | Year: 2011

This paper reports a new approach for the modeling of SCC crack growth in nuclear coolant environments. Many SCC tests have been carried out since 70's. Some of crack growth rate data was observed to be widely scattered. Previous studies have concentrated on the prediction of crack growth rate in a wide range of stress intensity factor. On the review of many laboratory test data, we propose a process of crack growth of SCC is the coupled phenomenon as a result of synergistic state transition between mechanical system that consisted of stress and material, and chemical system that consisted of material and environment. The coupled phenomenon is a two-stage reaction as results of state transitions of material which commonly exists in the both systems. Since in the case of new approach, several critical reactions could exist in the process of SCC crack growth, it could be explained that an observed data in laboratory test was fluctuated. Source


Chikazawa Y.,Japan Atomic Energy Agency | Kotake S.,Japan Atomic Power Company | Sawada S.,Hitachi - GE Nuclear Energy
Nuclear Engineering and Design | Year: 2011

Comparative evaluation of fast reactor pool/loop configurations in the JAEA feasibility study (FS) from 1996 to 2006 has been re-evaluated from the new point of view. In the FS, both pool and loop configurations have been investigated. The FS loop concept (FS-loop) is an advanced two-loop design which is being developed in the Fast Reactor Cycle Technology Development (FaCT) project as Japan Sodium-cooled Fast Reactor (JSFR). In this study, a brief description of the FS pool concept (FSpool) has been provided. The FS-pool pursues a compact reactor vessel structure using a flask shape intermediate heat exchanger. The original FS pool/loop comparison in 2000 concluded that material amounts of those systems were about similar between pool and loop configurations. However, the present comparison based on the reactor vessel structure and primary cooling system pointed out relatively good economic potential of the FS-loop concept, because differences in the secondary cooling systems of the pool/loop configurations are not essential from the viewpoint of pool/loop comparison. In addition, a rough estimation method of a reactor vessel diameter against power has been proposed and the result has been compared with RV diameters of various pool/loop concepts. The discussion on the reactor vessel diameter shows that the FS-pool reactor vessel is the most compact of the recent commercial reactor designs such as SPX-2, SNR-2, EFR, CDFR and BN-1600. This study points out better economic potential of FS-loop (present JSFR) as a result of competition with the most compact pool concept. © 2010 Elsevier B.V. All rights reserved. Source


Naganuma M.,Japan Atomic Energy Agency | Ogawa T.,Japan Atomic Energy Agency | Ohki S.,Japan Atomic Energy Agency | Mizuno T.,Japan Atomic Energy Agency | Kotake S.,Japan Atomic Power Company
Nuclear Technology | Year: 2010

In the Fast Reactor Cycle Technology Development (FaCT) project, a sodium-cooled fast reactor (SFR) with mixed-oxide (MOX) fuel and an SFR with metal fuel were selected as the primary and the secondary candidates, respectively, for the Japan Sodium-Cooled Fast Reactor (JSFR). The present study focuses on the effects of transuranium (TRU) composition in the design for the JSFR core with MOX fuel. In the transitional stage from light water reactor (LWR) to fast breeder reactor (FBR), there is the possibility for FBR fuel to have high minor actinide (MA) content due to the recycling of LWR spent fuel. High MA content affects core and fuel designs as follows: the neutronic reactivity characteristic changes; the linear power limit is reduced because of decreases of the melting point and thermal conductivity in the fuel; the gas plenum length is extended because of an increase in He gas generation. Thus, to evaluate the effects quantitatively, design studies for cores with two TRU compositions were conducted: an FBR multirecycle composition with ∼1 wt% (in heavy metal) of MA content and an LWR recycle composition for which 3 wt% of MA content was assumed as a tentative target. The results show that the change from the FBR multirecycle composition to the LWR recycle composition leads to a sodium void reactivity increase of 10%, a linear power limit decrease of 1 to 2%, and a gas plenum length increase of 5%. As a result, the effects of TRU composition on the core and fuel designs were revealed to be benign. Source


Yamano H.,Japan Atomic Energy Agency | Kubo S.,Japan Atomic Power Company | Kurisaka K.-I.,Japan Atomic Energy Agency | Shimakawa Y.,Mitsubishi Group | Sago H.,Mitsubishi Group
Nuclear Technology | Year: 2010

An advanced large-scale sodium-cooled fast reactor named JSFR adopts an innovative two-loop cooling system. This cooling system design raises major technological issues: hydraulic and structural integrity due to the increase in one-loop coolant flow rate, safety design against the break or failure in one-loop piping, and ensuring the reliability of the decay heat removal system (DHRS). The present paper describes the investigation of the piping structural integrity due to flow-induced vibration using a ¿-scale hot-leg piping test. The structural integrity of the hot-leg piping in the JSFR design has been confirmed by a flow-induced vibration analytical methodology, verified with the experimental data. Additional experimental results have revealed that hydraulic issues including gas entrainment and vortex cavitation could be prevented by some design measures. By applying appropriate safety design, the two-loop system has been confirmed to be valid against the break or failure in one-loop piping by a safety evaluation in this study. The DHRS with natural circulation is designed in conformity with the two-loop system by introducing adequate safety designs. In this paper, the validity of this DHRS is given by a probabilistic safety assessment and safety evaluation. Source

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